A nuclear reactor comprises a core of fissionable fuel which generates heat during fission. The heat is removed from the fuel core by the reactor coolant, i.e. water, which is contained in a reactor pressure vessel. Respective piping circuits carry the heated water or steam to the steam generators or turbines and carry circulated water or feedwater back to the vessel. Operating pressures and temperatures for the reactor pressure vessel are about 7 MPa and 288.degree. C. for a boiling water reactor (BWR), and about 15 MPa and 320.degree. C. for a pressurized water reactor (PWR). The materials used in both BWRs and PWRs must withstand various loading, environmental and radiation conditions. As used herein, the term "high-temperature water" means water having a temperature of about 150.degree. C. or greater, steam, or the condensate thereof.
Some of the materials exposed to high-temperature water include carbon steel, alloy steel, stainless steel, and nickel-based, cobalt-based and zirconium-based alloys. Despite careful selection and treatment of these materials for use in water reactors, corrosion occurs on the materials exposed to the high-temperature water. Such corrosion contributes to a variety of problems, e.g., stress corrosion cracking, crevice corrosion, erosion corrosion, sticking of pressure relief valves and buildup of the gamma radiation-emitting Co-60 isotope.
Stress corrosion cracking (SCC) is a known phenomenon occurring in reactor components, such as structural members, piping, fasteners, and welds, exposed to high-temperature water. As used herein, SCC refers to cracking propagated by static or dynamic tensile stressing in combination with corrosion at the crack tip. The reactor components are subject to a variety of stresses associated with, e.g., differences in thermal expansion, the operating pressure needed for the containment of the reactor cooling water, and other sources such as residual stress from welding, cold working and other asymmettic metal treatments. In addition, water chemistry, welding, heat treatment, and radiation can increase the susceptibility of metal in a component to SCC.
In particular, intergranular stress corrosion cracking (IGSCC) occurs in nuclear plant piping systems. Certain weld materials exposed to the reactor water environment are highly susceptible to IGSCC. The present methodology for addressing IGSCC-susceptible piping systems is to remove the entire piping system and replace it with new materials. This is an expensive approach as it involves special materials, requires disposal of old radioactive materials and is very labor intensive. Thus, there is a need for a system for upgrading/repairing existing piping systems without removal/replacement of the old piping.